Nuclear Systems Research Articles

Coupled neutronics, thermal-hydraulics, and fuel performance analysis of dispersion plate-type fuel assembly in a cohesive way

The nuclear reactor is a multi-physics and multiscale coupling system. The conditions of high temperature and high neutron flux in the reactor pose a significant challenge to the thermodynamic performance of the fuel. To evaluate the plate-type fuel's performance, a coupling calculation of neutron physics code OpenMC, fuel performance analysis code BEEs-Plates, and nuclear reactor system analysis code NUSAC has been implemented based on the open-source MOOSE platform. A novel and efficient method for mesh grids generation and data mapping between external codes and MOOSE was proposed to ensure accurate data exchange. This method can simplify the mesh grids generation process of external codes and accurately transfer data across different mesh grids. Finally, the coupling calculation of the JRR-3 M research reactor fuel assembly was carried out based on the above coupling code, and the neutronics parameters, thermal–hydraulic parameters, and fuel performance-related parameters of the fuel assembly were analyzed. The calculation results show that the maximum fuel burnup will reach 16.91% FIMA after 240 days. With the increase of fuel burnup, the peak power of the fuel assembly is significantly shifted towards both ends of the height direction. The VonMises stress, volume stress, creep strain, and displacement of the fuel will be considerable. Finally, the impact of the coupling calculation on the accuracy of the computational efficiency is preliminarily investigated.

Atlas of historical and proposed nuclear devices and power systems in the Earth-Moon System and wider Solar System

The 2010s witnessed a renewal in international interest of space operations extending outside geosynchronous orbit to encompass the cislunar environment and beyond with invigorated U.S., Chinese, and Russian lunar mission initiatives; planned commercial lunar projects; and coalescing international efforts to reach Mars and push deeper into the Solar System. With the space beyond the near-Earth orbital regimes likely to become competitive and congested in the coming decades, reliable and high-yield means of generating power and propulsion will be needed to enhance mission assurance and operational responsiveness. Nuclear systems represent a promising technology with a rich heritage and provenance of space-based applications to satisfy emerging requirements for electrical power and propulsive impulse for long-term human development of the Earth-Moon system and its planetary neighbors. In this work, an atlas for defining the varying historical or proposed uses of nuclear power in space operations and exploration is presented according to spatial zone within the Earth-Moon and wider Solar System. Depending on the spatial zone, nuclear power will be discussed with respect to applications such as electrical power generation, propulsion, resource extraction, and weapon testing and employment for the space control and planetary defense missions.

Highly efficient group separation of hexavalent actinides from lanthanides through a biphasic cooperative extraction system

The implementation of actinide (An) and lanthanide (Ln) group separation is one of the most critical tasks for the GANEX process to adapt nuclear fuel cycle in the Generation IV nuclear energy system. In this work, we proposed a biphasic cooperative extraction system to fulfill this aim, achieving selective extraction of hexavalent An to the organic phase by a hydrophobic extractant, N,N-di-(2-ethylhexyl)butylamide (D2EHBA), while masking trivalent or tetravalent Ln in the aqueous phase by a hydrophilic ligand, N,N,N’,N’-tetraethyldiglycolamide (TEDGA). Efficient group separation of actinides from lanthanides was achieved with remarkable single-stage An/Ln separation factors over 102. Moreover, the mechanism of An(VI)/Ln(III/IV) selectivity was analyzed by UV–Vis, FT-IR and MALDI-TOF-MS spectroscopic analyses, indicating more hydrophobic nature of An(VI)-D2EHBA complex than Ln(III/IV)-D2EHBA complex as well as much stronger masking effect of Ln(III/IV) than An(VI) by TEDGA contributing striking An/Ln separation.

Effects of cold rolling and heat treatment on microstructure and mechanical properties of 15Cr-15Ni ODS austenitic steel

Oxide dispersion strengthened (ODS) austenitic steels are considered as the promising candidate cladding materials application for advanced nuclear energy systems. However, the microstructure of ODS austenitic steels fabricated by mechanical alloying is usually not very homogeneous. Thermal mechanical treatment is an effective method for controlling microstructure of alloys. In this work, the effect of cold deformation and subsequent heat treatments on the improvement of microstructure homogeneous and mechanical property of a 15Cr-15Ni ODS austenitic steel was systemically investigated by means of electron backscatter diffraction, transmission electron microscopy and tensile test. Results show that the average grain size decreased and the distribution of grains was more homogeneous after optimized thermal mechanical treatment. The deformation of cold rolling and heat treatment temperatures shows obvious influence on the average size of dispersion particle, the proportion of high angle grain boundary, Schmidt factor and tensile property of 15Cr-15Ni ODS austenitic steel.

Effects of solutes on thermodynamic properties of (TMZrU)C (TM = Ta, Y) medium-entropy carbides: a first-principles study

High entropy carbide ceramics have garnered significant interest as a novel class of ultra-high temperature and superhard metallic materials. In the present work, a comparative investigation was conducted for the first time on the stability, mechanical, and thermodynamic properties of two medium entropy carbides (MECs), (TaZrU)C and (YZrU)C, using high-throughput first-principles calculations. Additionally, data from groups IV and V transition metal monocarbides were employed for comparison. The temperature-dependent thermodynamic properties, including bulk modulus (B), constant volume/constant pressure heat capacity (Cv/Cp), Gibbs free energy, volume, entropy, and thermal conductivity, were evaluated using the Debye-Gruneisen model. The results demonstrate that (TaZrU)C and (YZrU)C exhibit similar trends in their thermodynamic properties, with (YZrU)C displaying slightly superior performance as the temperature rises. This work provides valuable insights into the design of innovative high entropy fuels, holding significant implications for the advancement of MEC ceramic fuels in advanced nuclear power systems and nuclear thermal propulsion systems.

Roadblocks to Disarmament in the Nuclear Non-Proliferation Treaty System

Abstract For the last five decades, the states parties to the Nuclear Non-Proliferation Treaty (NPT) have come together every 5 years at Review Conferences to explore how they can advance the Treaty’s goals, including the pursuit of nuclear disarmament. Despite lengthy negotiations, the Review Conferences have failed to reach a consensus final document as many times as they have succeeded and even when a final document has been agreed it has contained few tangible gains for those working towards a nuclear weapon free world. This state of affairs played out again in 2022 when the NPT’s 10th Review Conference ended without agreement and with many states decrying the weakness of the disarmament provisions that had been under discussion. It has been common in the aftermath of Review Conferences for academics to consider the contemporary political issues that stymied progress. While such studies are important, it is also necessary to consider what deeper structural issues may be at play in preventing the goal of nuclear disarmament from being advanced in the NPT context. To this end, this article identifies and analyses three key structural issues with the NPT and NPT Review Conferences that create difficulties for those pursuing a nuclear disarmament agenda. First, it explores issues with the scope and content of Article VI and how it has been interpreted by the nuclear weapon states. Secondly, it looks at the difficulties that arise from the commitment to achieving consensus outcomes at the Review Conferences and, thirdly, it considers the limitations of who actually attends and contributes to Review Conferences.

Pyrocarbon Coating on Granular Al2O3 for HTGR-Type Power Reactor

Fourth-generation nuclear power systems are based on high-temperature gas-cooled reactors, in which the pebble fuel is the primary energy carrier. In this regard, applying protective pyrocarbon coatings on granulated fuel is an essential problem in ensuring the reliability of nuclear power plants. The article’s main idea is to research rational technological parameters of forming a pyrocarbon protective coating on the granules of a nuclear fuel model. For this purpose, granulated Al2O3 with the protective pyrocarbone coating was applied as a fuel model. The article’s aim is to study the effect of thermophysical parameters on applying a protective pyrocarbon coating on granulated Al2O3. During the experimental studies, thermal imaging of the pyrolysis process was used. The scientific novelty of the work is the equilibrium curves for the systems Al2O3:CH4, Al2O3:CH4:N2, and Al2O3:CH4:Ar. Their analysis allowed for evaluating rational thermochemical parameters of the pyrolysis process. As a result, rational thermophysical parameters of coating granulated Al2O3 with a pyrocarbon layer were evaluated, and the practical possibility of applying the pyrocarbon coating to granulated Al2O3 in the electrothermal fluidized bed was experimentally proven. It was shown that nitrogen does not significantly affect the target reaction product under a temperature of less than 1500 K. Also, the rational conditions for the pyrocarbon coating at a pressure of 0.1 MPa were realized at a temperature of 900–1500 K and using argon. Moreover, pyrocarbon was precipitated from hydrocarbon at 1073–1273 K. Overall, the need to add an inert gas for reducing the carbon black formation was proven to prevent a reduction of natural gas efficiency.

Benchmarking exercise of indigenous Monte Carlo neutron transport code PATMOC for hexagonal full core reactor geometry

An indigenous Monte Carlo neutron transport code, named PATMOC, has been designed and developed at Bhabha Atomic Research Centre (BARC) for reactor physics design applications. The code has been extensively tested for its algorithmic accuracy against numerous international numerical benchmarks, has also been validated on practical nuclear reactor systems like Advanced Heavy Water Reactor Critical Facility (AHWR-CF) and Apsara-U pool type research reactor located at BARC, and is now in regular use for day-to-day reactor physics studies. These reactor systems and most of the benchmark problems used for the verifications and validation of the code have square lattice arrangement in the core. In this paper, we present the results of verification of PATMOC code on whole-core reactor system with hexagonal lattice arrangement of fuel elements. In this exercise, we have analyzed the High Temperature engineering Test Reactor (HTTR) whole-core benchmark for integral and differential parameters at both lattice and core level. The results demonstrate the algorithmic accuracy of the code on hexagonal reactor systems also. This exercise will help make use of this code for reactor physics design studies also for the reactor systems with hexagonal arrangement.

Assessment of flow and heat transfer of triply periodic minimal surface based heat exchangers

An efficient intermediate heat exchanger is crucial for ensuring the safe operation, lifespan and energy efficiency of the advanced nuclear systems. Benefiting from the development of additive manufacturing technology, it is possible to design and optimize heat exchangers by introducing triply periodic minimal surface (TPMS) structures. With the excellent mechanical and thermal performance of TPMS, further improvements in the energy efficiency of advanced nuclear systems are achievable. In this study, I-WP surface, Primitive surface based heat exchangers and a printed circuit heat exchanger for accelerator driven subcritical systems are constructed in three dimensions. In order to obtain the potential enhancement of thermal performance of TPMS-based heat exchangers, the fluid flow and conjugate heat transfer characteristics in TPMS heat exchangers, especially for the specific working medium of lead-bismuth eutectic (LBE) are investigated. A parametric analysis is conducted on the key design variables including the solid volume fraction and hydraulic diameter. The results indicate that TPMS-based heat exchangers could achieve about 2–3 times the total heat transfer rate with half the volume of a printed circuit heat exchanger. The TPMS topologies contribute greater heat transfer enhancement on the Helium side than the LBE side, which helps to optimize the thermal resistance allocation in Helium-LBE heat exchangers. An increase of solid volume fraction enhances the convective heat transfer of Helium in I-WP heat exchanger, while homogenizing the local thermal performance. This study provides a database for the design and optimization of TPMS-based heat exchangers, which contributes to the sustainable future target.

Response and Mechanism of Coal Fine Production to Differential Fluid Action in the Baode Block, Ordos Basin

The Baode Block in the Ordos Basin is currently one of the most successfully developed and largest gas field of low–medium rank coal in China. However, the production of coal fine has affected the continuous and stable drainage and efficient development of this area. The special response and mechanism of differential fluid action during the drainage process is one of the scientific issues that must be faced to solve this production problem. In view of this, the evolution laws of a reservoir’s macro–micro physical characteristics under different fluid conditions (fluid pressure, salinity) have been revealed, and the response mechanism of coal fine migration-induced reservoir damage has been elucidated through a nuclear magnetic resonance online displacement system. The results indicated that pores at different scales exhibited varying patterns with increasing displacement pressure. The proportion of the mesopore and transition pore is not affected by salinity and is positively correlated with displacement pressure. When the salinity is between 3000 mg/L and 8000 mg/L, the proportion of macropore and micropore showed parabolic changes with increasing displacement pressure, and there was a lowest point. The evolution law of pore fractal dimension and permeability change rate under the action of different fluids jointly showed that there was an optimal salinity for the strongest reservoir sensitivity enhancement effect. The mechanical and chemical effects of fluid together determined the damage degree of coal reservoir induced by coal fine migration.

Design and optimization of intermediate heat exchanger for lead-bismuth cooled fast reactor coupled with supercritical carbon dioxide Brayton cycle

Printed circuit heat exchanger has become a new choice for the intermediate heat exchanger of the lead/lead–bismuth cooled fast reactor coupled with supercritical carbon dioxide nuclear power system because of its large heat transfer area density, high power density and ability to work at high temperature and high pressure. In this study, the channel dimensions of the intermediate heat exchanger were optimized to improve its thermal–hydraulic performance. An in-house code for the thermal–hydraulic design of printed circuit heat exchanger with lead–bismuth and supercritical carbon dioxide as working fluid was built. The relationship between the cold side pressure drop of intermediate heat exchanger and cycle efficiency of the nuclear power system is revealed. The power density and pressure drop of the cold side are selected as two optimization objectives, and the multi-objective optimization process is conducted with the non-dominated sorting genetic algorithms II. The optimal design results of the intermediate heat exchanger are obtained with Pareto-front method. The Pareto-front shows that when the heat exchange density increases, the pressure loss of the cold side also increases, but more quickly than the former. Finally, considering the design limitations, the intermediate heat exchanger core design scheme is selected by the auxiliary function which considering both capital cost and operating cost.

A novel transfer CNN with spatiotemporal input for accurate nuclear power fault diagnosis under different operating conditions

Deep Neural Network (DNN) models, recognized for their exceptional feature extraction and end-to-end diagnostic capabilities, have gained considerable attention from researchers in the realm of nuclear power fault diagnosis. However, existing deep-learning based diagnostic technologies present certain challenges when applied to practical engineering scenarios. A significant issue is the limited adaptability of data-based machine learning diagnostic methods to dynamic and evolving operational conditions, leading to inconsistent accuracy under different operating parameters. To solve this problem, a spatiotemporal Convolutional Neural Network (CNN) model is proposed. A novel input scheme, considering both spatial and temporal aspects, is specially designed for nuclear power plant fault diagnosis and MLP convolutional layer are used to improve the feature extraction capability of the model. Finally, the performance of the model is analyzed through a series of experiments based on simulation data and data visualization. The results show that the proposed method outperforms traditional CNNs in diagnosing nuclear power system faults with higher accuracy. Furthermore, by employing a transfer learning strategy, the model exhibits improved speed for cross-operating condition fault diagnosis.

Analysis of Kobayashi benchmark with indigenous Monte Carlo neutron transport code PATMOC

The solution of the neutron transport equation is the basic input for the reactor physics design of a nuclear reactor system. Because of the complexities of geometry and cross-section data, the neutron transport equation is generally solved using numerical methods. One of the difficulties in the solution using these methods concerns the accuracy of distribution of neutron flux in the system containing void regions in a highly absorbing medium. The difficulty arises because of the issue in the continuity of flux at the void and material interface. There is a sudden and large change in the flux at this interface. Kobayashi benchmarks are widely used problems for testing the ability and accuracy of reactor physics codes to compute neutron flux distribution on such systems. An indigenous Monte Carlo neutron transport code, named PATMOC, has been designed and developed at Bhabha Atomic Research Centre (BARC) for reactor physics design applications. The Kobayashi benchmarks have been used to test the PATMOC code to verify its accuracy in flux computation for systems with voids in-between high absorbing materials. Our results show that the PATMOC estimated values of flux distribution in the system compare very well with the reference results provided with the benchmark. In this paper we present the detailed results and analyses of this benchmark with PATMOC code.

Development of a one-dimensional system code for the analysis of downward air-water two-phase flow in large vertical pipes

In nuclear thermal-hydraulic system codes, most correlations used for vertical pipes, under downward two-phase flow, have been developed considering small pipes or pool systems. This suggests that there could be uncertainties in applying the correlations to accident scenarios involving large vertical pipes owing to the difference in the characteristics of two-phase flows, or flow conditions, between large and small pipes. In this study, we modified the Multi-dimensional Analysis of Reactor Safety KINS Standard (MARS-KS) code using correlations, such as the drift-flux model and two-phase multiplier, developed in a plant-scale air-inflow experiment conducted for a pipe of diameter 600 mm under downward two-phase flow. The results were then analyzed and compared with those based on previous correlations developed for small pipes and pool conditions. The modified code indicated a good estimation performance in two plant-scale experiments with large pipes. For the siphon-breaking experiment, the maximum errors in water flow for modified and original codes were 2.2% and 30.3%, respectively. For the air-inflow accident experiment, the original code could not predict the trend of frictional pressure gradient in two-phase flow as ⟨jg⟩/⟨j⟩ increased, while the modified MARS-KS code showed a good estimation performance of the gradient with maximum error of 3.5%.

Verification strategy for artificial intelligence components in nuclear plant instrumentation and control systems

The application of artificial intelligence (AI) systems in nuclear power plants has the potential to facilitate significant advancements in the industry. However, nuclear system engineers and regulators have expressed concerns regarding the reliability of AI components owing to a lack of documentation and appropriate verification methods that differ from those used for traditional software. Accordingly, this study proposes a systematic approach for the verification and validation of AI components based on nuclear software regulatory guidelines. The proposed strategy includes process verification and functional testing activities specifically tailored to AI components for instrumentation and control systems. This approach serves as a foundation for fostering understanding and agreement among system engineers, operating licensees, and regulatory bodies on the use of AI components in nuclear power plants.

Transient thermal–hydraulic analysis of the thermionic space reactor TOPAZ-II

Space nuclear reactor system is of great significance in aerospace applications such as deep-space exploitation and exploration, high-power earth-orbiting satellites. A transient reactor system analysis code, RESYS, is developed for advanced non-light water reactor safety analysis recently. In this paper, the model of the thermionic space nuclear reactor TOPAZ-II including the thermionic fuel elements, thermionic energy conversion system, reactor core, coolant loop and radiator is established with RESYS code. The point reactor equation with six-group delayed neutron is utilized to calculate the fission power. The reactivity effect of various structural components and the reactivity insertion effects of control drums are considered. The thermionic emission model of cesium thermionic converter is based on the phenomenological model established by Rasor. Then, the steady state operation and three transient accidents including reactivity insertion accident (RIA), partial loss flow accident (LOFA), load resistance shortcircuit accident (LRSCA)are simulated. In the steady-state condition, the output electric power is about 5.8 kW with thermoelectric conversion efficiency of 5.0%. The results of steady calculation agree well with the design values. And in the three transient accidents, the moderator introduces a large temperature reactivity feedback which makes the reactor fission power increasing. The results indicate that the TOPAZ-II reactor system has a sufficient safety margin during these accidents. In addition, the calculation results preliminarily demonstrate the capability of RESYS code in modeling and simulation of thermionic space reactor.

Effect of gas hydrate formation and dissociation on porous media structure with clay particles

Natural gas hydrates are a potential future energy modality with the advantages of clean burning and large resource reserves and have attracted worldwide attention. Natural gas hydrates formation and dissociation impact the skeletal structure and mechanical properties of porous media sediments. In addition, there is a synergistic effect between the migration of fine clay particles in porous media and hydrate behavior. In this study, methane hydrate was examined in-situ using a low-field nuclear magnetic resonance system in the presence of two suspensions of clay particles with different stability. The results showed that clay particles impacted methane hydrate formation and dissociation and the pore structure of porous media. Hydrate nucleated preferentially in large pores, causing them to split into smaller pores; the presence of clay particles, especially illite, improved the water conversion rate. The results indicated that water content in large pores increased after hydrate dissociation but was discontinuous among different pores. When illite was present, the distribution was more continuous; when montmorillonite was present, the water distribution of the large pores was similar to that of the original state. This work increases our understanding of the kinetics, water migration, and pore structure alteration of methane hydrate formation and dissociation in sediments with clay particles and provides support for the safe and efficient development of natural gas hydrates.

He-ion Irradiation Effects on the Microstructures and Mechanical Properties of the Ti-Zr-Hf-V-Ta Low-Activation High-Entropy Alloys.

High-entropy alloys (HEAs) have shown promising potential applications in advanced reactors due to the outstanding mechanical properties and irradiation tolerance at elevated temperatures. In this work, the novel low-activation Ti2ZrHfxV0.5Ta0.2 HEAs were designed and prepared to explore high-performance HEAs under irradiation. The microstructures and mechanical properties of the Ti2ZrHfxV0.5Ta0.2 HEAs before and after irradiation were investigated. The results showed that the unirradiated Ti2ZrHfxV0.5Ta0.2 HEAs displayed a single-phase BCC structure. The yield strength of the Ti2ZrHfxV0.5Ta0.2 HEAs increased gradually with the increase of Hf content without decreasing the plasticity at room and elevated temperatures. After irradiation, no obvious radiation-induced segregations or precipitations were found in the transmission electron microscope results of the representative Ti2ZrHfV0.5Ta0.2 HEA. The size and number density of the He bubbles in the Ti2ZrHfV0.5Ta0.2 HEA increased with the improvement of fluence at 1023 K. At the fluences of 1 × 1016 and 3 × 1016 ions/cm2, the irradiation hardening fractions of the Ti2ZrHfV0.5Ta0.2 HEA were 17.7% and 34.1%, respectively, which were lower than those of most reported conventional low-activation materials at similar He ion irradiation fluences. The Ti2ZrHfV0.5Ta0.2 HEA showed good comprehensive mechanical properties, structural stability, and irradiation hardening resistance at elevated temperatures, making it a promising structural material candidate for advanced nuclear energy systems.

Heat transfer and entropy optimization for unsteady MHD Casson fluid flow through a porous cylinder: Applications in nuclear reactors

Heat transfer and entropy generation are crucial considerations in the nuclear industry, where the safe and efficient transfer of heat is essential for the operation of nuclear reactors and other nuclear systems. Casson fluid is a useful tool in the nuclear industry for simulating the flow behavior of nuclear fuels and coolants, and for optimizing the design and operation of nuclear reactors. In view of this, the current investigation deals with the heat and fluid flow of unsteady Casson fluid in a circular pipe under the influence of magnetic field, internal heat generation, entropy generation and porous media. The governing equations have been simplified under suitable assumptions and nondimensional quantities. The simplified dimensionless governing equations have been solved using the method of separation of variables along with Bessel functions. It is concluded from the investigation that the temperature increases with time. The Casson fluid parameter raises the temperature and entropy generation. The temperature, entropy generation and Bejan number are the decreasing functions of the Prandtl number.

Measurement of the ^{181} Ta(n,gamma ) cross sections up to stellar s-process temperatures at the CSNS Back-n

The neutron capture cross section of ^{181} Ta is relevant to s-process of nuclear astrophysics, extraterrestrial samples analysis in planetary geology and new generation nuclear energy system design. The ^{181}Ta(n,gamma ) cross section had been measured between 1 eV and 800 keV at the back-streaming white neutron facility (Back-n) of China spallation neutron source(CSNS) using the time-of-flight (TOF) technique and hbox {C}_{6},hbox {D}_{6} liquid scintillator detectors. The experimental results are compared with the data of several evaluated libraries and previous experiments in the resolved and unresolved resonance region. Resonance parameters are extracted using the R-Matrix code SAMMY in the 1–700 eV region. The astrophysical Maxwell average cross section(MACS) from kT = 5 to 100 keV is calculated over a sufficiently wide range of neutron energies. For the characteristic thermal energy of an astrophysical site, at kT = 30keV the MACS value of ^{181}Ta is 834 ± 75 mb, which shows an obvious discrepancy with the Karlsruhe Astrophysical Database of Nucleosynthesis in Stars (KADoNiS) recommended value 766 ± 15 mb. The new measurements strongly constrain the MACS of ^{181}Ta(n,gamma ) reaction in the stellar s-process temperatures.